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Benchmarks and test facilities of VVER reactors
Šimek, Ondřej ; Foral, Štěpán (referee) ; Vojáčková, Jitka (advisor)
The aim of this bachelor’s thesis is to describe the topic of benchmarks and test facilities for pressurized water reactors of the eastern concept VVER type. The theoretical part introduces the VVER type reactors, their history, the development of separate generations, including fundamental differences and parameters of the VVER type reactors being used so far. Next two chapters of the theoretical part focus on nuclear safety and security, describing the uppermost Czech authorities, which are SÚJB, MAAE and NEA. Furthermore, many terms concerning nuclear safety are explained in these two chapters. The next chapter is focused on deterministic safety analyzes, their classification, methods and purposes. The aim of this part is also to explain the verification and validation of computing codes. The next chapter offers insight into test facilities which are crucial for evaluation and testing of nuclear devices and computing codes. The last chapter of the theoretical part focuses on the VVER type reactor benchmarks. The practical part of this thesis presents the conversion of a single AER Benchmark FCM_001 using neutron physical code PARCS. The results are compared to the results of CRONOS computing code.
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Depletion calculation of VVER 1000 reactor fuel using KENO code
Janošek, Radek ; Katovský, Karel (referee) ; Novotný, Filip (advisor)
The introduction to operational nuclear reactors focusing on light-water pressurized reactor VVER 1000 is in the beginning of this Master´s thesis. This thesis covers basic technology of VVER 1000 reactor with focus on reactor core and nuclear fuel TVSA-T. A significant part of the thesis deal with basic concepts of nuclear safety and its methods. The main goal is to create a model of VVER 1000 reactor, which can be used in nuclear burn-up calculations using KENO code. Therefore a part of this thesis deals with explanation of statistical Monte Carlo method and the KENO code.
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Approach to the nuclaer safety of the 3rd generation nuclear reactors
Pavlíček, Michal ; Kolář, Jaroslav (referee) ; Kolář, Jaroslav (referee) ; Matal, Oldřich (advisor)
The main target of the master´s thesis is reviewing the generation III nuclear reactors in term of the nuclear safety. At first we have to learn some theory of the nuclear safety in order to understand safety systems of the generation III nuclear reactors. Therefore the thesis is divided into two parts. Legislative and technical approaches to nuclear safety are mentioned in the first part. Regulatory bodies, whose task is to supervise nuclear safety in the nuclear power plants, belongs to the legislative approaches. There are defined terms such as defence in depth, redundancy, diversity, etc. There are mentioned methods to assessing nuclear safety – deterministic and probabilistic methods, especially probabilistic methods, for which a simple example is provided. There are also mentioned active and passive safety systems and their significance for nuclear safety and inherent safety too. There is an example of the function of the active and passive safety systems of the EDU nuclear power plant in conclusion of this issue. The second part deals with description of the selected nuclear reactors in context of the construction of the new units of nuclear power plant in Temelín. The nuclear reactors from companies, which applied for the public tender opened by ČEZ, a. s., for the construction of the ETE 3+4. Thus, the nuclear reactor MIR-1200 by ATOMSTROYEXPORT (Russian Federation), the nuclear reactor AP1000 by WESTINGHOUSE (USA) and the nuclear reactor EPR by AREVA (France) are taken into account . Comparison of the generation II and these generation III+ nuclear reactors necessarily belongs to this master´s thesis. These the generation III+ nuclear reactors are compared with the nuclear reactor VVER 440 (EDU) and in particular with the nuclear reactor VVER 1000, which is operated in the nuclear power plant Temelín. The final chapter contains generally appraisal of the whole problem.
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Accidents of light water reactors
Bejček, Patrik ; Novotný, Filip (referee) ; Foral, Štěpán (advisor)
This thesis deals with the description of primary and secondary circuit of nuclear power plant with reactor VVER-400. Moreover, this thesis describes nuclear safety, its particular factors, and also international nuclear event scale. It is followed by a description of companies playing a significant role in the field of nuclear safety and a close attention is paid to SÚJB. In a section dedicated to nuclear safety, the focus is on events which can occur in nuclear power plant. In addition, the accident Large Break LOCA is described in detail. The practical part deals with the description and instructions to simulator PCTran developed by Micro-Simulation Technology. In the last part, there is an analysis of selected parameters which are simulated in simulator PCTran, parameters simulated in simulator RELAP5 5 and also the calculation of the temperature parameters of fuel TVSA-T and their simulation.
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FACTORS LIMITING LIFE TIME OF NUCLEAR POWER PLANTS WITH PRESSURIZED-WATER REACTORS
Křivánek, Robert ; Kolat, Pavel (referee) ; Liszka, Ervin (referee) ; Fiedler, Jan (advisor)
The aim of the thesis is to analyze the state of preparedness of nuclear power plants (NPP) for long term operation (LTO) based on the IAEA SALTO (Safety Aspects of Long Term Operation) peer review service, analysis of the most significant failures, accidents and operational experience with type reactors PWR/VVER focusing on cases caused by equipment ageing and identification of major structures and components limiting life time of PWR/VVER-type nuclear power plants, and possible measures to ensure their required service life. Based on the results of the IAEA SALTO peer review service, an analysis of the main deficiencies and measures of NPPs in preparation for a safe LTO was performed, focusing on topics whose deeper knowledge is important for the future more precise determination of technical factors limiting the lifetime of NPPs. The main deficiencies and measures in the preparatory phase for LTO and the most important technical measures are summarized in chapter 4.5. The main deficiencies and the most important technical corrective measures in the area of ageing management of structures and components are discussed separately. The history of major failures and operational experience of nuclear power plants with PWR/VVER reactors from the point of view of ageing of structures and components is analyzed in chapter 6.2. The result is a statistic analysis of ageing-related events, an overview of the most significant PWR/VVER reactor failures with an impact on their service life, a statistical overview and discussion of the most important degradation mechanisms, and other important findings from the history of major failures and operational experience. Chapter 6.3 analyzes factors limiting the operation of nuclear power plants with PWR/VVER reactors with focus on structures and components potentially limiting the life of PWR/VVER reactors and possible measures to ensure their required life. In conclusion, the main reasons of permanent shut down of NPPs (actual and potential) for 40, 60 and 80 years of operation and the measures to ensure their required life are summarized.
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The concept for passive cooling of the VVER-1000 reactor
Lamoš, Pavel ; Suk, Ladislav (referee) ; Martinec, Jiří (advisor)
This thesis is focused on the design of passive cooling system for a nuclear reactor VVER- 1000.This type of reactor is located in the Czech Republic in the location of Nuclear power plant Temelín. The thesis states an overview of the different cooling systems of nuclear power plants. The thesis is focused on passive safety system especially on passive cooling system, so there was done an overview of currently used passive safety system. In the work is discussed nuclear safety and the maximum projected accident of VVER-1000, which is called LOCA accident. In the design part of the thesis was done thermal calculation of heat exchangers. Exchangers are designed as condensers with a natural flow, where cooling of system is provided by outside airflow in case an accident. The results are evaluated at the end of the thesis.
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Nuclear Power Plant Dukovany Tasks within an Electricity Grid of the Czech Republic
Prokop, Ondřej ; Řež,, Martina Malá, Centrum výzkumu (referee) ; Katovský, Karel (advisor)
The aim of the Bachelor’s thesis is to describe the role of The Nuclear Power Plant Dukovany (Dukovany NPP) in the Czech electricity grid. Increasing of the production of electricity from renewable energy sources in the electricity grid causes more frequently compliance of N-1 criterion and it decrease the reliability of the electricity grid. The risk of Blackout increases, which would cause large economic losses and danger to the life of human beings. Therefore, reliable, good controllable and safety sources of electricity are very important. The first part of the thesis describes the role of Dukovany NPP in the process of regulation of basic electrical parameters during the normal operation of electricity grid. The second part of the thesis deals with states and the function of Dukovany NPP after full or partial disintegration of the electricity grid. Final part analyzes the importance of Dukovany NPP in Czech electricity grid and future development Dukovany NPP with possibility of the construction of new nuclear units after the end of the operation in 2035. The Dukovany NPP is an important regulator of the reactive power. There isn’t other power plant in the southeastern part of the Czech Republic, which could significantly control reactive power. The power from Dukovany NPP, which is provided on ancillary services controlling the real power and frequency has been increasing in recent years. This is due to the increasing production of renewable sources. Units switch to island operation after partial disintegration of electricity grid. If frequency exceeds the limit, the units will switch to it is own consumption. In this stage of operation at own consumption can Dukovany NPP operated without the time limit. Dukovany NPP may participate in restoring the electricity grid, based on dispatch control instruct. If switching to own consumption is not successful and back up sources and emergency’s dieselgenerator are unavailable, the power supply has to be restored as soon as possible. There are six independent solutions to restore the power supply. Next option of restore power supply will be the AAC network after the completion of construction in 2014. The AAC network will be resistant to the extreme weather and will be place it in the area of the power station, thereby Dukovany NPP will be more independent on external network.
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Nuclear power plants and their safety
Novotný, Petr ; Köbölová, Klaudia (referee) ; Milčák, Pavel (advisor)
The bachelor thesis deals with the current state of nuclear energetics. It describes the development of nuclear power plants to date and specifically describes the most commonly used types of nuclear reactors. Most of the reactors currently in operation are second generation reactors. However, a number of third-generation reactors with improved technological and safety parameters are already under construction. Improved safety lies in the improvement of active safety systems, but also in the increased use of passive safety systems. The paper concludes with a comparison of specific solutions from the most important nuclear reactor manufacturers in the world.
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Aplikace SW Apros pro vytvoření modelu bezpečnostního systému JE Dukovany
DOLEŽAL, David
In nuclear power plants, we encounter a number of safety systems that complement, replace or assist each other and thus reliably perform their safety functions. A number of advanced tools are currently used to develop and test these safety systems, which must meet not only industry standards but also legislative regulations and directives. These tools include the very powerful industrial simulation program Apros 6 Nuclear, which was developed directly for use in nuclear power plants, and is therefore able to help meet the requirements of nuclear power plants in the field of nuclear safety. Within the theoretical part of this work, the issues of safety of industrial units, safety of nuclear facilities and demonstration of the Apros 6 program itself, including a description of the work in this program, are presented. This therefore means acquaintance with the most important terms used in nuclear safety, with the structure and classification of safety systems used in nuclear power plants and with the basic principles used in the field of nuclear safety. The last theoretical chapter introduces the Apros 6 program itself, the basic tasks that can be performed in it and which were used in the practical part to create a model structure. After getting acquainted with the theoretical contexts, the practical part of the work emphasizes the characteristics of a specific safety system of the primary circuit of Dukovany NPP and the implementation of the model of this safety system in the program Apros 6 Nuclear. The content of the model is an introduction and description of all the most important parts together with their settings and location.
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